I

LA-5506-PR‘regress Report UC-48

Issued: May 1974

CIC-14 REPORT COLLECTION

ic3 REPFKxNJCTiON●COPY

Preparation and Evaluation of

Medical-Grade Plutonium-238 Fuels

July 1,1971- June 30,1973

Compiled by

L. ]. MuhN3

}

)

Q’/●10s + alamosscientific laboratory

of the University of CaliforniaLOS ALAMOS, NEW MEXICO 87544

uNITED STATES

ATOMIC ENERGY cOMMISSION

CONTRACT W-740 S-ENG. 3S

ABOUT THIS REPORTThis official electronic version was created by scanning the best available paper or microfiche copy of the original report at a 300 dpi resolution. Original color illustrations appear as black and white images.

For additional information or comments, contact: Library Without Walls Project Los Alamos National Laboratory Research LibraryLos Alamos, NM 87544 Phone: (505)667-4448 E-mail: lwwp@lanl.gov

This report was prepared as an eccount of work sponsor~ by the United

States Government, Neither the Unitad Statas nor the United Statas AtomicEnergy Commission, nor any of their employees, nor any of their contrac-

tors, subcontractors, or their employees, makes any warranty, express or im-plied, or assumes any legal liability or responsibility for the accuracy, com-pleteness or usefulness of any information, apparatus, product or processdis-closed, or represents that its use would not infringe privately ownad rights.

Work funded by the US AEC Division ofApplied Technology. The program wasassigned to the Division of Biology and En-vironmental Research in July 1973.

Printed in the United States of America. Available fromNational Technical Information Service

U.S. Department of Commerce5285 Port Royal Road

Springfield, Virginia 22151Price: Printed Copy $4.00 Microfiche $1.45

,?.

CONTENTS

ACKNOWLEDGMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..v

ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...1

1, INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...1

II. REVIEWOF238PUFUELPROPERTIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...2

III. PLUTONIUM-238 FUEL DEVELOPMENT FACILITY . . . . . . . . . . . . . . . . . . . . . . 3

IV. PROCESS DEVELOPMENT STUDIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...4A. Fabrication of Artificial Heati Fuel Cylinder for Vented Heat Source . . . . . . . . 4B. Preparationof238PuOz16byOxygen-Isotopic Exchange . . . . . . . . . . . . . . . . . . . 8

v. PREPARATION AND PROPERTIES OF PROCESS INTERMEDIATES . . . . . . . ...10A. ElectrorefinedMetaland23%u0216Powder . . . . . . . . . . . . . . . . . . . . . . . . ...10B. (a,n)Contributionfrom 170and180in238pu0216 . . . . . . . . . . . . . . . . . . . . ...13

VI. PREPARATIoNANDPRoPERTIEsoF338puoJ6PRoDucTs . . . . . . . . . . . . ...13A. One-Watt PeUete . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...13B. Fifty-Watt Cylinders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...13C. Thirty-Three-Watt Cylinders for VentedHeat Sources . . . . . . . . . . . . . . . . . . .14D. Twenty-Nine-Watt Heat Sources for Implantation Studies . . . . . . . . . . . . . . . . .16

VII. ENCAPSULATION OF NONVENTED HEAT SOURCES . ., . . . . . . . . . . . . . ...16-~ A. Quality Control Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...17-— e. B. Encapsulation ..,,.... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...17~

,~%. I.Fifty-WattS ounces . . . . . . . . . . . . . . . . . . . . . ..O.O.. . . . . . . .0.00..178===~~%! 2, Twenty-Nine-Watt Sources . . . . . . . . . . . . . . . . . . . . . . . . . 00 . .00000.18

?~m my a. Ta-lOWWeld DevelopmentProgr= . . . . . . . . . . . . . . . . . . . . . . ...18l+m b. Encapsulation of29-WCYbders . . . . . . . . . . . . . .. . . . . . . . . . . . ...18:E~

3==O==~rT/lll.HELI~ RELEASEFROM238PUOJ6WEL CYLINDER ..,.. . . . . . . . . . . . . . ...20~~,~= m A. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...20‘-m’ B. Experimental . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...22

~m ;...— l. Apparatus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ., . . . . . . . ...22

a, Steady-State Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...22b. Temperature-Excurrnon Testig . . . . . . . . . . . . . . . . . . . . . . . . . . ...23

2. Fuel Cylinder, . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . ...243. Procedure . . . . . . . . . . . . . . . . . .. OOOOOO. OO.O. . . ...00.0.000.024

a, Steady-State Testing...,.. . . . . . . . . . . . . . . .’ . . . . . . . . . . . . . ...24b. Temperature-Excureion Testig . . . . . . . . . . . . . . . . . . . . . . . . . . ...25

C. Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...25

...111

IX. RADON RELEASE FROM%uO#6FUEL CYLINDERS . . . . . . . . . . . . . . . . . ...27A. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...27B. Experimental . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...29

l. Apparatus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...292. Fuel Cylinder . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...293. Procedure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...29

C. Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...31

x. FUEL CHARACTERIZATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...31A. NonddructiveAnalysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...31

l. Neutron Counting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...312. Gamma Spectrometry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...32

a. Meaeurementof170and180inqu0216 . . . . . . . . . . . . . . . . . . . . ..32b. Determinationof=~h . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...32c. Minicomputer DataReductionof Gamma-RaySpectra . . . . . . . . . . ...32

3. Calorimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...32B. Chemiccd Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...32

1. MassSpectrometricAnalysisof170and 180in%%0216 . . . . . . . . . . . . ...322.MeasurementofO/MRatiosin2%%02 . . . . . . . . . . . . . . . . . . . . . . . . ...33

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...33

APPENDIX . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...35I. SI Units . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...35II. Standardized Nomenclature System forHeat-Source Identification . . . . , . . . . .35

iv

ACKNOWLEDGMENTS

Thin work was performed principally by CMB-1 and CMB-11personnel. However, the assistance of several other LASL groupsis gratefully acknowledged. The following groups, in particular,rendered valuable contributions.

CNC-4 – made and provided all the H2016 used in thisproject.

CMB-6 — fabricated all of the ceramic ware and much of the

specialized equipment; designed and tested the

dies and punches fox the fuel form for the ventedheat source.

The work was done under the direction of R. D. Baker, CMBDivision Leader. Principal investigators were L. J.Mullins and G.R. Waterbury. The following individuals contributed to the workreported here: A. J. Beaumont, R. G. Bryan, J. Bubernak, M. R.Conner, B. A. Dye, C. L. Foxx, N. L. Koski, G. M. Matlack, T. K.Marshall, L. J. Mullins, A. K. Murdock, M. E. Smith, A. N.Morgan, and R. L. Nance.

v

PREPARATION AND EVALUATION OF

MEDICAL-GRADE PLUTONIUM-238 FUELS

JULY 1, 1971- JUNE 30, 1973

Compiled by

L. J. Mullins

ABSTRACT

Work has continued on the preparation and properties of238PuOZ 16 fuel forms for the artificial heartmedical-grade

program. Procedures were developed for fabricating the 33-Wfuel cylinder for the artificial heart vented capsule. Threesources were fabricated and delivexed to Mound Laboratory(MRC ) for encapsulation in vented capsules. One SO-W and five29-W nonvented heat sources were fabricated for animal implan-tation studies.

Studies of helium and radon release rates from medical-gradefuel forms have been initiated and are in progress.

The characterization of ‘6Pu fuel forms by nondestructivetechniques and chemical analysis has continued.

Most of the investigations discussed are of the continuingtype. Results and conclusions described may therefore bechanged or augmented as the work continues. Publishedreferences to results cited in the report should not be madewithout obtaining explicit permission to do so from the personsin charge of the work.

I. INTRODUCTION

The principal objectives of the medical-grade fuelprogram axe (1) to prepare and evaluate medical-grade 2%u fuels for an artificial heart device, (2) tofabricate and characterize heat-source capsules (upto 50 W) for dosimetxy, systems studies, and ixnplan-tation studies, (3) to develop diagnostic proceduresfor evaluating ‘6Pu fuel forms, (4) to establish thecompatibility of medical-grade fuel forms withselected container materials, and (5) to measureproperties of medical-grade fuel forms important tothe design and safety of the he_@ source.

The 2MPu requirements (N69 g) for an artificialheart or circulatory-assist device demand a fuel hav-

ing minimal radiation properties. Prepaxing andevaluating potential 238Pu fuel forms have led to thedevelopment of four fuel compositions having radia-tion properties approximately equal to the irreduci-ble minimum values for 23%. The fuel forms areelectxorefined metal, 238Pu-3 at.Yo Ga, %uNIS, and~6Pu@6. Theoretical and experimental studies ofthese fuels led to the conclusion that 2WPU0 216is thepreferred composition for high-temperature( w870”C) application in the artificial heart program.This fuel is prepared as a pressed and sinteredceramic cylinder. Procedures have been developedfor preparing and characterizing sources varying insize from 1 to 50 W.

1

The work done from the start of this program inJuly 1967 to June 30, 1971, was reported in detail inLA-4940.1 LA-5506-PR covers the period July 1, 1971,to June 30, 1973.

II. REVIEW OF’ 238pu FUEL PROPERTIES

Penetrating radiations from 23*Pu fuels having theisotopic compositions shown in Table I are derivedprimarily from

1. spontaneous fission neutrons,2. neutrons produced by (atn) reactions in im-

purities of low atomic number,. 3. fast fission neutrona,

4. photons from the decay of ‘%J,5. photons from other plutonium isotopes and their

daughters, and6. photons that result from alpha particle reactions

with light element impurities.An irreducible neutron emission rate is associated

with the spontaneous fission of ‘% and -u. Forthe spontaneous fission of 23%, the average numberof neutrons per fission ia 2.33(+0.08)2 and the half-life ia 5.0(+0.6) x 1010 yr.3 These values give acalculated value of 2586(+400) n/s-g ~%. Resultsobtained in the Los Alamos Scientific Laboratory(LASL) program give a value of 2785(+55) n/s-g238Pu for neutron emission resulting from the spon-taneous fission of ‘8Pu.1j4

The neutrons resulting from (arn) reactions ob-viously depend on the particular elementi present

TABLE I

PLUTONIUM ISOTOPIC COMPOSITION OFCONVENTIONAL AND MEDICAL-GRADE

‘Pu FUELS(Typical Lots)

Abundance, wt%Pu Isotope Conventional Medical-Grade

236 1 x 10+

TABLE H

POTENTIAL MEDICAL-GRADE

%% FUEL COMPOSITIONS

TABLE 111

COMPARISON OF THE DOSE RATES OF

%% FUEL FORMS a~b~c

MeltingComposition Temp (C)

Pu 640Pu-3Ga 670Pu -10 Sc 750PU4 Zr 800(Pu Fez + Fe) 1165Pu Pt* 1475Pu C12 1650Pu NIS 2200Pu 0216 2400

Power Density (mega-watts per cubic meter)~b

7.2

7.16.65.93.93.25.45.64.2

%steulations based on a 2%% isotopic composition of 80 at.%sad a fuel density of 90% of rhcoretical.

bFor the most part, S1units are used in this report as given in theMemic Practice Guide, E380-72, American Society for Testing andMaterials. These units are compared to the units of Ref. 1 in theAppendix, Table A-I. Exceptions to S1 usage are noted when used.

power density than the nitride or carbide, ita ex-cellent chemical stability, high melting point, andease of synthesis and fabrication make it the primefuel candidate for most medical applications.23*u02 has been studied for several years as a fast-reactor fuel and much information has beengenerated on its chemical and physical properties.

Fuel forms that have been prepared, characteriz-ed, and evaluated are

Elemental 238Pu,

238Pu -3 at. YO Ga,238pu02nat ,

2WU0216,ZS8PUNIIat,and238puNlSo

The gamma and neutron dose rates of these fuelsare compared in Table III. The gamma dose rate ofthe metal 100 mm from the center of a 9.52- by 9.52-mm cylinder contained in a 0.76-mm-thick tantalumcontainer is 0.25 R/h-kg 23~u. The neutron dose rateis 0.31 rem/h-kg 238Pu. Dose rates for the Pu-3 at. YOGa alloy are identical to the metal. The neutron doserate of PuNnat is the same as the metal; however, thegamma dose rate is approximately doubled as the

14N yielding an excitedresult of an (a,p) reaction onatate of 170. Replacement of 14N by 1~ gives acompound having a gamma dose rate only 10%higher than the metal. (The slightly higher gamma

(95.2-rnrn by 95.2-mm Cylinders)

Gamma NeutronComposition R/h-kg 2% retn/h-kg2%

Pu 0.25 0.31Pu -3 at.% Ga 0.25 0.31Pu Nna[ 0.54Pu NIS

0.310.28 0.31

PU02 nat 0.36 1.50PU0216 0.29 0.34

at% isotopic composition, 80 wt%2%%.bDose rates in air through 0.76 mm T%10 cm from the center ofthe source.

CAI1fuels prepared from electro refined metal.

dose rate of the nitride is due primarily to ita lowerdensity.) Both the neutron and gamma dose rates of238Pu02n& are significantly higher than the metal.However, 23@u0216 has dose rates comparable tothe metal. Thus, the fuel forms having low neutronand gamma dose rates are 238pu, 238Pu.3 at.yo Ga,2QSpuN15, -d 238PU02 16. Insofm as radiation is

concerned, all of these materials would be suitablefor implantable heat sources required to powerpacemakers and artificial heart devices.

III. PLUTONIUM-238 FUEL DEVELOPMENTFACILITY

The medical-grade 238Pu fuel development facilityis shown in Fig. 1. This facility consists ,of three linesof glove boxes. The line of hydrogenous-shieldedboxes on the right- is used for aqueous and drychemistry operations which can be performed in air-filled glove boxes. In addition to the air boxes, thisline also has a small inert atmosphere glove box,argon-flush type, in which 23*Pu metal can behandled. The center line consists of two large inertglove boxes, recirculating-argon type with gaspurification units. One of these boxes is used forpreparing and casting electrorefined metal andalloys and for preparing 2%h0 }6 powders. Theother box is used to fabricate ceramics, and the mill-ing, pressing, and aintering operations are done in

3

.

this box. The line of boxes on the left is presently usedfor the temporary encapsulation of heat sources.

The fuel development facility is used to preparemedical-grade fuel forms, to recycle processresidues, and for process development studies. (Flowsheets for the medical-grade fuel form preparationwere presented in Ref. 1.)

IV. PROCESS DEVELOPMENT STUDIES

A. Fabrication of Artificial Heart FuelCylinder for Vented Heat Source

A drawing of the 238Pu0216 cylinder for the TRW*-designed 33.1- W (thermal) artificial heart heatsource is given in Fig. 2. Fuel cylinders prior to thisdesign were all right-circular cylinders of simplegeometry and had a height-to-diameter ratio ofN 1.0.1 The cylinder shown in Fig. 2 has a height-to-diameter ratio of 2.1 and has a cavity to accept apressure relief device (PRD); all edges arechamfered. To use existing equipment and minimizedevelopment costs, ceramic fabrication proceduresutilizing drypressing facilities were developed.

The following approach was used in developingthe ceramic fabrication procedures.

1. Dies and punches were designed and made fordry-pressing a “green” plutonia cylinder in onepiece. The green body contained all the geometrical

*TRW Systems Group, Redondo Beach, CA.

2.5 mm mln x

+t-4.83 mm diem, mln

90°55°

%1.

6.35 mm diem, mln

450\mwm

VIA 35.81 mmmox

.27 mm min x 45°~

w

17.Omm diemmox

I. I

Fig. 2.Artificial heart fuel cylinder, TRW design.

features of the designed cylinder except for the topchamfer in the PRD cavity. The dies were designedassuming a minimum shrinkage of 5~0 during thesintering process. The green cylinder and die con-figuration are shown in Fig. 3. The top edge of thePRD cavity is chamfered after the cylinder issintered. This operation is easily accomplished byhand with a conical diamond countersink.

2. Thoria cylinders were made using these dies todemonstrate the practicability of the fabricationdesign.

3. Three238Pu0216 cylinders, 80 at.% enrichment,were made to establish fabrication procedures beforefabricating three medical-grade 33-W cylinders.

When the above work was concluded, ceramicfabrication problems still remained, The foremostproblems were obtaining integral “green” bodiesand “crack-free” sintered cylinders. A series of ex-periments was therefore initiated to ,examine theseproblems more fully and to establish the .reproducibility of both product and process. Tencylinders were prepared in this study, and allcylinders were made from nominal 80 at. YO238Pu “ menrichment. The same weight of oxide powder wasused in these preparations as would be used in thepreparation of a 33. 1-W medical-grade cylinder. Thefirst five cylinders, 38 through 42, were made fromthe same batch of 238Pu0 216 powder. This batch of

4

es

cylinder

nch

DFig. 3.

Cylinder and die configuration for pressing ar-ti~cial heart fuel cylinder.

powder was prepared from old SO-W sources madefor earlier program commitments. The procedure Iwas as follows.

1. The SO-W cylinder was broken up by milling for30 tin in a SPEX Mill.

2. The resulting powder was heated to 760”C in anatmosphere of (Ar + H2016) and held at thistemperature for approximately 100 h. The powderwas then heated to 1540”C, held at this temperaturefor 8 h, and then cooled to ambient temperature.

3. The powder was ball milled for 18 h in auranium-molybdenum ball mill.

4. The powders (395-g total weight) were mixed ina large, shallow, water-cooled pan.

5. The cylinders were prepared by the proceduregiven in Ref. 1, p. 59.

The composition of the blended powder ia given inTable IV. The powder was made from cylinders andcontrol pellets from the preparation of cylinders 10,11, 13, and 24. (Note: Only a small amount of 10 wasavailable for use because most of this fuel was used toprepare the TRW developmental cylinder 34.)

TABLE IV

COMPOSITION OF POWDER FOR29-W CYLINDERS

Serial Numbers38 through 42

Total Weight of Powder 395.12 g

CompositionSO-WCylinder

SerialNo. Weight Fraction

10 0.060511 0.314713 0.309624 0.3154

~hesc cylinders were formerly identified ss 50-3,50-4, 50-5, snd 50-6. The nomenclature system inpresent use is defined in the Appendix.

Three cylinders (38, 39, and 40) and the controlpellets were pressed during one work period andwere then sintered by the procedure given in Ref. 1,footnote (b), p. 60. Cylinders 41 and 42 were pressedduring a later work period. As was the case with theearlier 33-W TRW cylinders, some difficulty was en-countered in obtaining integral green bodies duringthe pressing operation. (For preparations in whichthe pressing was unsuccessful, the carbowax binderwas allowed to vaporize, the cylinder was rescreen-ed, and the pressing operation was repeated.) Twopressings were required for 38 and three for 40. Thefirst pressing was successful for 39,41, and 42. (Note:The fuel ejection equipment and procedures weremodified after fabricating 40. These modificationsresulted in obtaining integral green pellets on thefirst pressing attempt.)

The physical dimensions of the five einteredcylinders are given in Table V. The same weight ofoxide feed was used in each pressing. The differencein weighta of the sintered bodies ia due to varyingprocess losses. The average weight was 74. 19( *0.05)g. Assuming a power density of 0.3914 kW/kg fuel,the average power of the five cylinders should be29.04 ( +cO.02) W. Thus the preparation of sources ofconstant wattage should pose no problem if thespecific power of the feed material is known ac-curately. The average length of the cylinders was34.80 (*O. 15) mm. A comparison of the maximum

5

TABLE V

PHYSICAL DIMENSIONS OF 29-W23A0216 CYLINDERS

Cylinder Serial Number38 39 40 41 42a 47b— — — —

Wt, g fuel 74.20 74.11 74.16 74.18 74.30 73.83Dimensions, mm (specification - length 35.81 max, diam17.02 max)

Length 34.67 34.80 34.59 35.05 34.92 34.52Diam

Max 16.71 16.69 16.69 16.76 16.69 16.66Min 16.66 16.61 16.66 16.71 16.64 16.64

Cavity for PRD, mm (specification diam 4.83 rein, depth6.60 rein)

Diam 5.08 5.05 5.05 5.11 5.08 5.08Depth 6.86 6.81 6.86 6.96 6.93 6.83

‘A piece of the cylinder wall, weighing 1.9 g, broke out at themidpoint of 42 during the sintering cycle.

bcylinder 47 prepared by recycling 42 through ceramic fabrica-tion.

and minimum diameters of Table V shows that verylittle “hourglaseing” occurs in the aintering of thesecylinders. The average maximum diameter was 16.71(+0.02) mm and the average minimum diameter was16.6 (*0.02) mm. The average dimensions for thePRD cavity were 5.08 (*0.02) mm by 6.88 (* O.OS)

9 mm deep. The physical appearance of cylinders 38through 41 was excellent., A piece of the cylinderwall of 42 weigh@g 1.9 g broke out at midpoint dur-ing the sintering process. Subsequent radiographicexamination of thq other cylinders revealed nodefects in 38, 39, and 41. At least one Kansversecrack and one oblique crack were detected in 40.Cylinder 42 was reprocessed through ceramicfabrication (milling, screening, pressing, and einter-ing) to produce a ceramic of excellent physicalappearance. Subsequent radiographic inspection,however, revealed a crack about ~dway.along-the ----””length of the cylinder. This cylinder ia designated aa47.

The second group of cylinders, 43 and 44, wasprepared from Lot LAS 14 238Pu0216 powder whichwas prepared 18 months before ceramic processing.After removal from storage in an inert atmosphere,the material was subjected to the followingoperations.

6

1. Oxygen isotopic exchange at 750”C,2. 6-h heating at 1540”C, and3. 18-h ball milling.

Cylinders 43 and 44 were pressed three months afteroperation 3, above. No difficulty was experiencedwith the pressing operation; integral pellets were ob-tained on the first attempt Sintering of 43 and 44,however, resulted in low-density cylinders that didnot meet dimensional specifications. The diametersand lengths were too large. See Table VI. The densi-ty of the control pellets for these cylinders was only81% of theoretical as compared to the usual 87 to88%. Subsequent analysis of the particle size of thepowder used for preparing 43 and 44 showed asignificant difference from the usual ball-milledpowder. The particle size distribution of the 43 and 44powder is compared to the powder for preparations38 through 42 and the powder for preparation 47, asshown in Fig. 4. (The latter two powders gave high-deneity cylinders and are typical of those measuredon the usual ball-milled powder.’) The 43 and 44powder contains much larger particles than either ofthe other two powders and its distribution curve ismuch flatter. The larger size of the 43 and 44 powdermay be due to sintering of the powder during the 3-month storage of the ball-milled powder.

TABLE VI

PHYSICAL DIMENSIONS OF 29-W‘~0216 CYLINDERS

Serial Numbers43 through 46

43 44 45 46

Wt (g) 74.08 74.07 74.33 74.17Dimensions, mm (specification - length 35.81 max, diam

,17,02 ).. ~~Length 36.32 36.07 34.52 35.23Diarrs

Max 17.35 17.32 16.64 16.51Min 17.32 17.27 16.66 16.51

Cavity for PRD, mm (specification - diam 4.83 rein, depth6.60 rein)

Diam 5.28 5.31 5.08 5.11Depth 7.11 7.16 6.83 6.91

.

.

.

g I#so —

g ~. _&

540 —v&n-m –

47~

I I I I I I I I I I I I I I I I 1~00Z46810_12 141618

Size ( tO-6m)

Fig. 4.Particle size distribution of milled powders,preparations 38 through 44 and 47.

The control pellets for 43 and 44 were broken up ina SPEX mill and reprocessed through operations (2)and (3), above. The purpose of these operations wasto resize the feed powder and to release storedhelium gas, which might also affect pellet densities.This powder was then pressed into two small pelletsand sintered. The resultant densities were 88% of

theoretical. The particle size distribution for this feedpowder is shown in Fig. 5. Cylinders 43 and 44 werethen broken up and reprocessed through steps (1)through (3). The resulting powder, see Fig. 5 for par-ticle size distribution, was pressed into cylinders 45and 46. No difficulty was encountered in obtainingintegral green units on the first attempt, as shown inTable VII which summarizes the pressing ofcylinders 38 through 47. Sintering of the pelletsresulted in cylinders well within the dimensionalspecifications shown in Table VI.

The reproducibility of product weights is importantto meeting the wattage specifications. The weight ofproduct feed is adjusted to the desired amount before

g I I I I I I I I I I I I I I I+ 80

\ 4;@so—wE~40 -

&n 20 –

CmVrdpellet

=

.~oO 2 4 6 8 10 !2 14 16 18

I I I I I I I I I I I I

Size (10-6 m) “

.

Fig. S.Particle size distribution of resized control pelletpo waler and resizedpowder for preparations 4S and46.

TABLE VII

NUMBER OF PRESSINGS REQUIRED TOPRODUCE INTEGRAL “GREEN” CYLINDER

Cylinder8erisl Number

38394041424344454647

No. ofRessings

2131111111

adding carbowax (Ref. 1, p. 59). Small losses are en-countered in the subsequent operations which in-clude loading the die, pressing and extracting thefuel, sintering the cylinder, and, finally, chamferingthe top edge of the PRD cavity. The materialbalances for fabricating cylinders 38 through 47 aregiven in Table VIII. The average weight of the finaleintered and chamfered pellets was 74.14 ( AO.10) g.

The following conclusions can be drawn fromthese experiments.

a. Artificial heart heat sources of the design shownin Fig. 2 can be fabricated routinely by dry-pressingand sintering techniques.

b. The product weight can be controlled to betterthan 0.2 g. Thus the power output of a 33-W sourcecould be controlled to 0.1 W.

c. Powders should be pressed as soon as possibleafter preparation;

The remaining problem areas in fabricatingcylinders of the TRW design are

1. to fabricate “crack-free” cylinders on a routinebasis. (Approximately half of the pellets have con-tained cracks detectable by radiography.)

2. to eliminate trichloroethylene as a solvent forcarbowax, which is used as a pressing and sinteringaid. (Tnchloroethylene is decomposed by the heatand radiation from 23%1 and releases a corrosive gasor vapor. Since this gas is generated in an inert glovebox, its residence time in the box is relatively longand corrosion of equipment and work areas takesplace. Thus, replacement of trichloroethylene by aninocuous solvent is desirable.)

7

serialNo.

383944 ?424344454647

TABLE VIII

MATERIAL BALANCES FOR FABRICATION

OF CYLINDERS

SerialNumbers

38 through 47

Wt OxideFeed (g)

74.531274.518574.516974.503274.539274.505874.503074.526574.533874.502

W SinteredPellet (g)

74.215774.125574.180274.205074.347074.112574.041274.359074.190773.8595

B. Preparationof238Pu0216 by Oxygen-IsotopicExchange

Preparing 23%%0216by the reaction

2~u0216 + H20nd238pu02n~ + H2016 —

was first reported by Porter and Thompsons whoprepared powder having neutron emission rates aslow as 5500 n/s-g 23%u. Severed lots of 238pu0 216

powders have been prepared in the presentprogram1~6 by passing argon gas saturated withH2016 over plutonia at 700 to 750”C. The neutronemission rates of these powders were 300 to 1500 n/s-

9 238Pu higher than the measured rate for oxidesprepared from electrorefined metal. Oxygen isotopicexchange was complete in all cases and the higherrates were due to chemical impurities in the oxidefeed.

Only a limited effort was devoted to the exchangeprocess during the present reporting period. This ef-fort was devoted to

1. fabricating and evaluating a vertical columnthat allows H2 016 and a carrier gas (helium) to flowdown through a bed of 239u02. The pressure in theapparatus can be lowered to 1 x 103 Pa to increaseH2016 transport rate and thus increase the exchangerate.

2. improving the chemical purity of the oxide feedmaterial. The present effort was limited to the

Wt ChamferedPellet (z)

74.197074.106574.158874.175874.297574.090474.025274.331774.171773.8314

ProcessingLoss (g)

0.3340.4120.3580.3270.2420.3930.1950.1950.3620.671

volatilization of impurities from Savannah River ox-ide feed at 900°C.

The equipment for the isotopic exchange ex-periments is shown in Fig. 6. The reaction tube isplatinum-lined nickel with a platinum frit close to thebottom to support the bed of 2WPU0 z powder. Thesmall tube used has a capacity of -40 g of powder, Aquartz bubbler contains the H2 016. All H 20 in theoff-gas is trapped in two series-connected liquid N 2-cooled Pyrex trapa. A Welch Model 1397 vacuumpump was adequate to handle the flows used. All in-ner metallic surfaces between the bubbler and theplatinum in the reaction tube are gold plated toeliminate any possible reaction with H2 016.

Two experiments were done in the above equip-ment. In the first one, as-received Savannah RiverOxide (SRO), 238Pu02nd, Lot HBO 907545, wasused as feed material. The chemical analysis of thismaterial has already been reported. 1 In the secondrun, the same lot of oxide was heated in air at 900°Cfor 10 h before introducing it into the isotopic-exchange equipment.

The neutron emission rate of the oxide wasmonitored in situ during the experiments. Althoughthese measurements were semiquantitative, they didindicate when exchange was complete as evidenced

by a leveling-off of the neutron count. The operatingpressure above the bed of 10 g of oxide was 4 x 103 –9 x ld Pa depending on helium flow through theHz 016. The H2 016 flow rate through the oxide wae

8

.

.

rcIw.,011.o.., L’.”o~”1 :,,SO.,! 1:.T“k

IoaA)

Fig. 6.Oxygen isotopic exchange apparatus.

-1.6 g/h. The results of the neutron monitoring areshown in Fig. 7 where the relative neutron count rateis plotted as a function of time, (Relative neutroncount rate = 100 x [counts/rein attime(t) /counts/rein at time 0], In both experiments,approximately 5 h were required to reach completeexchange ao evidenced by a “constant” neutroncount rate. The experiments were continued for atotal time of 12 h at 800°C. When the experimentswere completed, samples of the powder were encap-sulated and evaluated. Table IX gives tbe neutronemission rates for the oxide feed and the 2*u0 216products. The neutron emission rates of the oxidefeed powders were 18600 and 16500 n/s-g 23*u.The neutron emission rate of the products were 3850and 3740 n/s-g ‘*u. The lower emission rate of thefeed for Experiment 2 was primarily due tovolatilization of fluorine by the 900°C ignition in air.However, as shown in Table X, this pretreatment ofpowder did not remove fluorine completely from thenatural oxide, Future volatilizations of impurities willbe done in high vacuum at higher temperatures.Table X also shows a definite pickup of nitrogenwhich was submquently largely removed duringisotopic exchange. An apparent increase of sodiumis indicuted by these results, but the precision of themeasurements must be improved before definite con-

.clueions can be drawn. (The results given in Table Xwere obtained by gamma epectrometry, which, atpresent,. is used to estimate relative amounts of im-.purities.) The oxygen isotopic analyses of the

100I I I I I I I I I

90 ~ Experiment IA——–A Experiment 2

z: 70 —

o

c 600L: 50z

\

~ 40 — \>.-: 30 —

z

u 20 — 4‘kA+A--a--A-- .

L--l

o~01234567 S9 10 II 12

Hours of H20’6 Admission

Fig. 7.Re[ative neutron counl rate as a function of time,oxygen-isotopic exchange experiments.

TABLE IX

NEUTRON EMISSION RATES OF OXIDEFEEDS AND PRODUCTS, ISOTOPIC

EXCHANGE EXPERIMENTS

(n/s-kg%%) x 10-3Expt 1 Expt 2

%hIOz ‘at feed 18600 16500‘8Pu0216 product 3850 3740

TABLE X

ANALYSIS OF LIGHT ELEMENT IMPURITIESIN IGNITED %%02nat and ‘SPU0216

Ratio, g Impuriw in Oxide/g Impurity in S. R. OxideIgnited S. R. 238p@ 16 2Wpu0216

tmpurity S. R. Oxide Oxide Expt 1 Expt 2

N 1.00 1.6 0.1F

0.11.00 0.3 0.3 0.3

Na 1.00 1.1 1.1 1.1cl 1.00 0 0.2 0

9

‘Hz 016 water and the ‘*u0216 products are givenin Table XI. These results show that oxygen-isotopicexchange was com lete. The neutron and gamma

&dose rates of the ~u0216 powders are given inTable XII. The neutron dose rates were 0.39 and 0.38R/h-kg ‘*u and the gamma dose rates were 0.91and 0.84 R/h-kg ~6Pu at 100 mm.

The resulta are compared with those obtained on238Pu0216 powders prepared from electrorefinedmetal and show that the product from Isotopic-Exchange Experiment 2 has a neutron emission rate360 n/s-g 23*Pu higher than the average ratemeasured on nine lots of powder. 1’4The gamma doserate of the Experiment 2 product is approximatelytwice that of the electrorefined-metal oxide powders.The high gamma dose rates are due primarily to236Pu daughters.

V. PREPARATION AND PROPERTIES OFPROCESS INTERMEDIATES

A. Electrorefined Metal and‘8Pu026 Powder

The two principal intermediates in the fabricationof medical-grade products are electrorefined metaland ~8’PuOz 16 powder. Procedures for theirpreparation were given earlierl with results forelectrorefining runs through 100-29 and for powderpreparations through LAS 13. An additional nineelectrorefining runs were made during the presentreporting period to produce a total of 606 g ofproduct. Approximately 420 g of this material had a23@u isotopic content of 80 at. Yoand the balance was80 at.% material. The chemical purity of the metalwas comparable to the product from earlier runs; 1however, to conserve ZS8PU,samples were not takenin every case. The neutron emission rates of productmetal from several runs are given in Table XIII. Theaverage emission rate for the nine 1- to 2-g sampleswas 3080 (+30) n/s-g ‘*u. This rate is comparableto that reported, earlier, 13095 (+46) for 18 samplescontaining 1- to 2-g plutonium. The chemical purityof a typical lot of electrorefined metal is given inTable XIV.

Few, if any, impurities are introduced into the‘VUO ~6 powder. Theproduct in preparing

chemical purity of a typical oxide is given in TableXV. A further indication of the purity of the oxidepowders is given by the gamma spectrum which candetect trace amounts of fluorine, sodium, chlorine,nitrogen, oxygen, phosphorus, and aluminum. Theimpurity elements detected in the gamma spectrumof powders are given in Table XVI. Most of the lots

TABLE XI

OXYGEN ISOTOPIC ANALYSIS OFHz 016 AND%%0216

Description 170 , ppma 180, ppma

H2016 8~b 15b

%%02 16, Expt 1 80C 1Oc

~02 16, Expt 2 70C 10C

‘g oxygen isotope/1000-kg oxygen.

bMeasuredby msss spectrometry.

cMcasuredby gamma spectrometry.

TABLE XII

DOSE IUiTES OF 2%02 16 PRODUCTS,ISOTOPIC EXCHANGE EXPERIMENTS

Dose Rates at 100 mma Expt 1 Expt 2

Neutron, rem/h-kg %% 0.39 0.38Gamma, R/h-kg%U 0.91 0.84

‘Measured in air through 7.d-mm Ta on 2Z samples.

showed no detectable impurities. The only elementdetected that would produce (a,n) neutrons issodium, which was found in lot LAS 17. Spectro-chemical analysis of this powder gave a value of 50ppm for the sodium concentration, which shows thatsodium can be detected at very low concentrationsby gamma-ray spectrometry. This method can alsobe used for estimating 170 and 1%3 in 2~u0 ~6.Oxygen isotopic analyses for several lots of powderby gamma-ray spectrometry is given in Table XVII.The 180 values are consistent with the 180concentration measured on H2016 used in thepreparations. The 170 values are higher than the 170measured on H2 016. Most of the lotsofH2016 used inthe preparations contained -75 ppm 170 (Ref. 4).

Work is continuing on the use of gamma-ray spec-trometry for the nondestructive quantitative deter-mination of 170, l@, and light element impurities(see Sec. X). The neutron emission rates of the eightlots of powder prepared during the present reportingperiod are given in Table XVIII. The averageneutron emission rate of the eight lots was 3375 (+45)n/s-g 2-u. This compares to an earlier value of 3378

.

.

.

10

RunNo.

100-30

100-31

100-35

100-36

100-42100-45

——— ___

TABLE X111

NEUTRON EMISSION RATES OF ELECTROREFINED METAL

SampleNo.

382385408409461467463464469524528529

.—— —— -

Pu (g)

1.61a5.09b1.52a1.66b1.94b

14.081.84b2.07b

17.211.601.681.65

Isotopic Comp~ 23%

89.8789.8789.9289.9290.4290.4290.42

90.4290.4279.81

79.8179.81

Neutron Emission Rate,(n/see-kg %%) x 10-3

311531253120310530403124305030303110310530803090

%mple from top of casting.

b%mple from bottom of casting.

TABLE XIV

CHEMICAL PURITY OF A TYPICAL LOT OF ELECTROREFINED ‘8Pu METAL, RUN 100-45

Element e

Li

\

TABLE XV

CHEMICAL PURITY OF A TYPICAL LOT OF23%%0216POWDER, LOT LAS 16

Element

LiBeBNaMg

AlSiKCaTi

vCrMnFeco

NaCuZnRbSr

ppma

B. (arn) Contribution from 170 and 180 in238pu(-j216

Most of the previousl’S’7 estimates of the (a,n)contribution from 160 and 170 in 23?Pu0 ~n”t and238Pu0216 were based on measurements made byRussian workers of the neutron yield from poloniumnitrate solutions.8 The conclusion drawn from thisstudy was that the (a,n) cross section for 170 wasapproximately 10% of that for 180. More recentexperiments by workers at Lawrence LivermoreLaboratory (LLL)9 give data on the cross sections ofthe (a,n) reaction on 170 and 180 for alpha particleenergies from 5 to 12.5 MeV. The ratio of 170 to 180cross section at 4.6 MeV is 0.41 showing that moreneutrons can be expected from (a,n) on 170 thanindicated by the measurements made on polonium.Thus for 23?PuO}6 containing 10 ppm 180 and 100ppm 170, the contribution from 180 and 170 wouldbe -250 n/s-g 238Pu* using the LLL cross sectionsand -100 n/s-g 238Pu* using the earlier Russiancross sections. As shown above, the average increasein neutron emission rate in converting electrorefinedmetal to 238Pu0216 powder is -275 n/s-g 2~u. Ourresults appear to be in good agreement with the LLLvalues.

VI. PREPARATION AND PROPERTIES OFZS8PU0216 PRODUCTS

A. One-Watt Pellets

The major emphasis in fuel preparations has beenon the fabrication and evaluation of 29-, 30-, 33-, and50-W sources for dosimetry experiments, animal im-plantations, and systems studies. However, smallsources, -l-W, were prepared for otherlaboratories for studies not connected with the heartprogram. The properties of the pellets prepared forMonsanto Research Corporation and LawrenceLivermore Laboratory are given in Table XIX. These

‘8Pu0 z16powder having anpellets were made fromisotopic composition of 86.90 wtYo ‘*u and 0.25 ppm2%%. The dimensions were 6.9 m’m diam by -8.5mm tall and the power was 1.3 W. The averageneutron emission rate of the pellets was 3250 n/s-g238Pu, which is somewhat lower than the rate,usually

*In making these calcukrtions a value of 1.18 n/s-g23qu was used for the (atn) contribution of naturaloxide,l which contains oxygen having an isotopiccontent of 99.759 at. YO 160.

observed on pellets of this size. The same lot of238PU0216 powder was used to fabricate 0.65-Wpellets for Mound Laboratory. The dimensions ofthese pellets were 4.8 mm diam by -4.9 mm tall. Theproperties of the pellets are given in Table XX. Theaverage neutron emission rate was 3310 n/s-g 2*u.The chemical purity of the pellets described in TableXIX and XX is given in Table XXI.

One-watt 238Pu0216 pellets, having an isotopiccomposition of 89.95 wtYo 23*Pu and O.18-ppm 236Pu,were made for Donald W. Douglas Laboratory. Theproperties of these pellets are given in Table XXII.The average neutron emission rate was 3350n/s-g 23%?I, which compares favorably to a value of3390 n/s-g 23*Pu for the oxide powder (Lot LAS 17)used in the preparation.

B. Fifty-Watt CylindersI

The preparation of 50-W sources 8, 9, 10, 11, and13 was reported in Ref. 1. Two additional 50-Wsources were prepared during this reporting period.The properties of these cylinders are given in TableXXIII. The neutron emission rates of the cylinderswere 3550 and 3591 n/s-g ‘8Pu. It should be notedthat the neutron multiplication factor in theserelatively large sources is approximately 6%. Themeasured rates on the 1-W conkol pellets were 3320n/s-g ‘*u for cylinder 24 and 3410 n/s-g 23%,1 forcylinder 25. Both cylinders were prepared from oxidepowder made from electrorefined metal. Thechemical purities of the metal, powder, milledpowder, and sintered cylinder are compared inTable XXIV. The analyses are fairly typical of theprocess used for the SO-W cylinders. The elec-trorefined metal is the purest material of the four

z~uo 216 powder comparesitems listed. Thefavorably to the metal. The ball milling and screen-ing operations introduce small amounts of impurities,some of which are removed in the subsequent sinter-ing operation. Plutonium isotopic analyses forcylinders 24 and 25 are given in Table XXV and theO/M ratios are given in Table XXVI. The oxides areetoichiometric Pu02.m within the limits of thechemical analyses used. (The O/M ratios were bas-ed on chemiccd analyses for oxygen and plutonium.A more precise method is now available; see Sec. Xbelow.)

13

Wt of fuel pellet (g)Wt of % (g)d

Pu isotopic compositionWt% -u isotopeppm ~ isotope

DimensionsDiarn, mmHeight, mm

Density (Mg fuel/m3 )Power (W)

Neutron emission ratenis-capsule(n/s-kg %)x 10_3

Dose rates at 10 cmNeutron remlh-kg ‘%%Gamma R/h-kg %e

TABLE XIX

PROPERTIES OF 1.3-W mPu0216 PELLETS FOR

MONSANTO RESEARCH CORPORATIONAND

LAWRENCE LIVERMORE I.ABORATORY

Pellet Serial Number

86.900.25

6.918.69

9.631.34

75253185

0.32—

3.0822.329

86.900.25

6.918.66

9.491.32

76503285

0.33-.

Slcp,b 1@ .- –.”

3.1232.364

19a

3.1522.382

86.900.25

6.918.89

9.451.35 “

77903270

0.33--

-72. 23’

3.0742.343

86.300.21

6.918.33

9.831.33

76203255

0.330.25

aPellets for Monsanto Research Grporation.

bA serial number was not assigned to this pellet. The sample number is therefore used for identification.

cPellets for Lswrence Livcrmore Laboratory.‘Value based on calorimetry and Pu isotopic analysis.

‘Measured in air through 0.076-cm Ta.

C. Thirty-Three-Watt Cylinders for VentedHeat Sources

Three 33-W fuel cylinders were made for encap-sulation by Mound Laboratory in the artificial heartvented fuel capsule. The dimensions and geometry ofthis 33-W cylinder are given in Sec. 11.A and in Fig.2. Because of the tight time schedule for delivery ofthe cylinders to Mound Laboratory, the cylinders,designated as 35, 36, and 37, were fabricated beforecompleting the ceramic fabrication developmentprogram (see Sec. IV.A, above). Three experimentswere done with ‘~u0216, 80 wtYo *u, beforefabricating the 33-W cylinders, 90 wt% ‘Pu. These

experiments demonstrated the

3.0752.348

86.300.21

6.918.36

9.841.33

76703265

0.330.25

feasibility off&icating the cylinder as one unit, since integral,but cracked, cylinders were obtained in two of thethree experiments. The procedures developed inthese experiments were used to fabricate the 33-Wcylinders.

The fabrication of 3S proceeded smoothly. An in-tegral green pellet was readily obtained and waseintered to meet dimensional specifications. Con-siderable difficulty was experienced in obtaining in-tegral green bodies for 36 and 37. Three attempts topress 36 powder resulted in laminated cylinders withthe break occurring at the bottom of the cylinder.Similar difficulties were encountered with 37, but

.

.

14

TABLE XX

PROPERTIES OF 0.65-W%@ 16PELLETS FOR MOUND LABORATORY

Pellet Serial Number14 15 16 311a 312= 313= ‘- ‘---—— .— —. 17

Wt of fuel pellet (g) 1.551 1.519 1.503 1.327 1.434 1.521Wt of % (g)b 1.171 1.147 1.134 1.002 1.083 1.148

Pu isotopic compositionWt% ~ isotope 86.9o 86.90 86.90 86.90 86.90 86.90ppm ‘Pu isotope 0.25 0.25 0.25 0.25 0.25 0.25

DimensionsDiam (mm) 4.83 4.88 4.88 4.88 4.85 4.88Height (mm) 8.84 8.64 8.64 7.92 8.64 8.92

Density (Mg PU02 /m3 ) 9.55 9.41 9.33 9.04 9.00 9.14Power (W) 9.664 0.650 0.643 0.568 0.614 0.651

Neutron emission ratenkeapsule 3920 3815 3785 3305 3450 3815

(n/s-kg -u) x 1~3 3345 3325 3335 3300 3185 3325—.——————.

1.4331.081

86.900.25

4.888.61

8.940.613

36403370

515-

1.3831.044

86.900.25

4.838.39

9.030.592

34453300

aA serial number was not assigned to this pellet. The sample number is therefore used for identification.

bValue based on calorimetry and Fu isotopic analysis.

only one pressing waa attempted. The powders for 36 To ensure geometrical epecificationa, a fuel accep-and 37 were then aintered at 1530°C using the stan-dard aintering cyclel and were remilled for 18 h.These powders were successfully pressed into greencylinders uing a new set of die punches. Thechemical purity of the three cylinders* is given inTable XXVII. The higher-than-usual values foraluminum, silicon, chromium, and iron are due to theexcessive handling and processing of the powdersduring ceramic fabrication. Virtually all of the 2UUpresent in the fuel is due to decay of ‘%% after fuelpurification.

The isotopic composition of the three 33-Wcylinders is given in Table XXVIII. The 238Pu isotopiccomposition was greater than 90 wtYo and the 23*uconcentration was less than 0.20 ppm. Thespecifications and properties of the 33-W cylindersare given in Table XXIX. The average neutron ernia-sion rate of the three cylinders waa 3460 n/s-g 23*u.

——— ——_—

*Chemical and isotopic analyses are done on thecontrol pellets that accompany all sourcefabrications. See Ref. 1, p. 59, for procedures.

tance gauge designed and fabricated by TRW waaused (TRW Drawing G41 1268, “Fuel Gauge Assy.,AHHS). Each of the three cylinders fit readily intothis gauge. The physical appearance of the cylinderswas good; however, subsequent radiographic ex-amination revealed one prominent transverse crackat the midlength of 35 and a crack at the upperchamfer on 36. Radiographic examination also in-dicated either a low-density area ox a possible crackabout 20 mm from the upper end of 36. There wereno radiographically detectable defects in 37.

The cylinders were packaged for shipment asshown in drawing CMB-1 1-9868, September 15, 1972.An earlier shipment of one of the 80 at.% develop-ment cylinders, 34, showed that this method of ship-ping did not result in darnage to the fuel cylinder.Shipment of the three cylinders to Mound Laboratorywas ‘accompanied by a LASL escort. Subsequentradiographic inspection of the sources at MoundLaboratory showed no apparent shipping damage.Calorimetric measurements at Mound Laboratorygave values of 33.042, 32.881, and 33.085 W, for 35,36, and 37. The TRW design specification is 33.1

15

TABLE XXI

CHEMICAL PURITY OF 1-W PELLET

Element

LiBeB

NaMgAlSiKCaTivCrMnFecoNiCuZn

ppma

TABLE XXII

PROPERTIES OF 1-W 23%%02 lb PELLETS FOR DONALD W. DOUGLAS LABORATORY

tipsule Number

26 27 28 400a 29 402a 403a 404a 30 31—— —

Wt of fuel pellet (g) 2.263 2.269 2.280Wt of ‘8PU (g)b 1.784 1.784 1.789

2.2571.775

2.2771.791

2.2821.795

2.244 2.250 2.2881.765 1.770 1.752

2.2641.782

89.95

0.18

7.98

4.55

9.981.010

5920

3320

Pu isotopic compositionWt% “%% isotope 89.95 89.95 89.95ppm 2~u isotope 0.18 0.18 0.18

89.95

0.1889.95

0.1889.95

0.18

89.95 89.95 89.950.18 0.18 0.18

DimensionsDiam (mm) 7.98 7.98 7.98Height (mm) 4.50 4.47 4.52

7.954.62

7.98

4.477.984.47

7.98 7.95 7.984.52 4.50 4.39

Density (Mg fuel/m3 ) 10.1 10.2 10.1Power (W) 1.011 1.011 1.014

10.21.015

10.21.017

— 10.1 10.11.000 1.003 0.993

--

1.006

Neutron emission ratenfs - capsule 5845 5930 5950 5940 5935 5995(nIs-kg U8PU) x 103 3275 3325 3325 3365 3355 3420

.—---.—. —n— —aA serial number was not assigned to this pellet. The sample number is therefore used for identification.

6005

338060253365

6035

3365

bValue based on calorimetry and Pu isotopic analysis.

The following sources were fabricated and source, product index sheet, and packaging anddelivered during the current reporting period.

One SO-W 238PuOZ 16 source to Battelle, PNL forimplantation studies,

Four 29-W ‘8Pu0216 sources to Battelle, PNLfor implantation studies, and ~

One 29-W 23@u0216 source to CornellUniversity for implantation stu~es.

shipping.

B. Encapsulation

1. Fifty-Watt Sources. Three SO-W 23%0 216cylinders, 13, 24, and 25, were encapsulated asshown in LASL Drawing 26Y-74527 (7-24-72) and Ex-perimental Plan 0820-EP- 1, Revision O. This capsuleconsists of three containers, an inner tantalum liner(0.51-mm wall thickness), a Ta-10W strengthmember (1.52-mm wall thickness), and a Pt-20Rhouter container (0.51 -mm wall thickness). Theclosure welds in the tantalum liner and Pt-20Rh cap-sules are cap welds and the closure weld in the Ta-10W capsule is a circumferential weld. The capsuleis shown in Fig. 10.

Considerable difficulty was encountered in the cir-cumferential closure weld of the Ta-10W strengthmember. Pressurization of the capsule during TIGwelding resulted in radical thinning of the wallthickness in two of the three capsules welded. Thethinning was detected visually in 5OLEX1O24 and the”

. .A. Quality Control Program

A quality control program was initiated in FY 1972for fabricating heat sources in the present program atLASL. This program is defined in document 0820-

Qc-p-l “LASL Quality Control Program for Difionof Applied Technology Products (Program 0820,Preparation and Evaluation of Pu-238 RadioisotopicHeat Sources), March 31, 1972.” Fabrication of the50- W heat sources is defined in Experimental Plan0820-EP-2. The experimental plans cover the fabrica-tion of the fuel, material procurement and accep-tance, fabrication, and preparation of capsule com-ponents, assembly, and “fabrication of the heat

17

TABLE XXIII

PROPERTIES OF SO-W23%%.102‘6 CYLINDERS

SerialNumbers

24 and 2524 25

Wt of fuel pellet (g Pu)Wt of *%U (g)

Oxygen/Pu Ratio

Pu isotopic composition

Wt% ‘8Pu isotope

ppm ‘Pu isotope

Dimensions

Diarn (mm)

Height (mm)

Density (Mg fuel/m3)Power (W)Power Density (W/cm3 )Neutron emission rate

(n/s-kg Pu) x 163(n/s-kg 23%) x 103

Dose rates at 10 cmNeutron, rem/h-kg %%Gamma, R/h-kg %hs

126.5189.21a

1.99

79.650.64

25.5025.48

9.7350.57

3.89

28273550

0.36o.2ob

125.0188.3 la

2.03

80.140.64

25.2225.15

9.9550.07

3.99

28783591

0.36o.19b

aCalculated from calorimetry and Pu mass spectrometry.

bMeasured through 0.051-cm Ta and O.OSI-cmstainless steel.

attempted repair resulted in a “blow-out.” The thin-ning in 5OLEX1O13 was detected by radiography.The welding of 5OLPX1O25 was satisfactory; theminimum weld thickness of the Ta-10W capsule inthis case was =83% of the wall thickness. No difficuhty was encountered with any of the tantalum liners orPt-20Rh capsule weldings; all welds passed inspec-tion. Heat sources 5OLEX1OI3 and 5OLEX1O24 wererejected because of the Ta- 10W welds. Heat source5OLPX1O25 was accepted and was shipped toBattelle, PNL, for implantation studies.

2. Twenty-Nine-Watt Sources. The capsuledesign for the SO-W sources was used for encap-sulating 29- W sources for implantation studies.However, the poor results obtained in welding theTa-10W capsules, above, pointed out the obviousneed for a weld development program beforeproceeding to the 29-W sources. An experimental

series of Ta-10W closure welds was therefore in-itiated to determine adequate welding parameters.

a. Ta- 10 W Weld Development Program. Cost con-siderations led to using test weld capsules fabricatedfrom two pieces of 25.4-mm-diam Ta-10W tubing asshown in Fig. 11. An inner tantalum liner (not shown)was used to approximate the clearances and voidvolumes of the heat-source capsule shown in Fig. 10.

Approximately 24 experiments were made usingvarious procedures and welding parameters. Thefrequency of success was encouraging with a venteddouble overlap scheme. This method uses a 0.25-mmtungsten heat shield between the Ta-10W and tan-talum in the weld area. A tantalum shim is also usedto ensure centering the liner. The proceduredeveloped is given below. Prior to its use with the 29-W sources, it was tried successfully on a full-scale en-capsulation of a Th02 cylinder and a full-scaleencapsulation of a 30-W cylinder (H/D = 1).

b. Encapsulation of 29-W Cylinders. Five 29-Wcylinders, 38, 39, 40, 41, and 47, have been encap-sulated as per Quality Control Program 0820-QCP-1,Revision O, and Experimental Plan 0820-EP-2. Theconfiguration of the capsule is given in Fig. 12. Allencapsulations were successful; cdl heat sourceswere accepted. The procedure for welding the Ta-10W capsule is given below.

Procedures for Ta-10W” Encapsulations of 29-WSources.

All part numbers refer to LASL Drawing 26Y-74997.

Procedures preceding welding are as follows.1. Vent the capaule, part 4, at 90° intervale.2. Lap parts 3 and 4 to gain intimate contact and

then clean as per Quality Control ProceduresPP-C-1, Rev. 1, paragraphs 4-6.

3. The source in its decontaminated and leak-checked tantalum liner is placed in Ta-10W,part 4, along with tantalum felt, tantalumshim, and tungsten strip (parts 14, 5, and 15,respectively).

4, Place the Ta-10W lid, part 3, on top andclamp the assembly in the welding fixturechuck. Adjust to eliminate inn-out of morethan 0.05 mm for part 4 and then align the lidusing a fixture.

Welding Procedures for Ta-10W.1. After a 50% Ar-50Y0 He atmosphere has been

established, tack weld the lid on at 120° inter-vals using 120 A.

2. Remove the alignment fixture, re-establiah a50% Ar-50% He atmosphere and perform

.

18

Element

TABLE XXIV

CHEMICAL PURITY OF SO-WCYLINDERS24, 25, AND PROCESS INTERMEDIATES

IMPURITY CONCENTRATION, ppm

Serial Number 24 Serial Number 25Milled Milled

Metal

LiBeBcNaMgAlSiKCaTivCrMnFecoNiCuZnRbSrY

. ZrMoCdSnCsBaLaHf

PbBiNp

TABLE XXV

PLUTONIUM ISOTOPIC COMPOSITION

Cylinders 24 and 25

wt%~POMass No. 24 25

236 0.64 X Id 0.64 X Id238 79.65 80.14239 16.29 15.91240 2.77 2.70241 0.94 0.90242 0.35 0.35

__________________ag of isotope/O.100-kgPu.

Weld No. 1 as shown in Fig. 13 at about 6 x104 Pa.

3. After 1/2 h cooling, perform Weld No. 2 asshown in Fig. 13.

4. Cool 1/2 h before breaking the inert at-mosphere and then leak check.

The actuaJ TIG welding parameters used for thefive 29-W Ta-10W capsules vaned alightly, but aretypified in Fig. 13.

VIII. HELIUM RELEASE FROM 23*u0 216FUEL CYLINDER

A. Introduction

The alpha decay of ‘%% continuously generateshelium within the ceramic fuel body. The rate andmanner of helium release from the oxide pellet in-fluence the internal pressures occurring within aheat-source capsule and dictate the helium flow rate

TABLE XXVI

OXYGEN TO METAL RATIO

Cylinderserial

Number Oxygen/MetaI

24 1.9825 2.03

TABLE XXVII

CHEMICAL PURITY OF 33-W CYLINDERS

Element

Li

Be

B

Na

Mg

AlSiKCaTivCrMnFecoNiCuZnRbSrYZrMoCdSnCsBaLaHfRePbBimlh237NP

2S2U

Serial Numbers 35, 36, and 37

Impurity (ppm)

35 36 37

TABLE XXVIII

PLUTONIUM ISOTOPIC COMPOSITIONOF 33-W CYLINDERS

Pu Weight Percent(g PU isotope/O.100-kgl%).

Isotope 39 36n 37a

236 0.14 X104 0.17 xl~4 0.18 X ld.238 90.14 90.24 90.04

239 9.22 9.13 9.30

240 0.595 0.589 0.614

241 0.030 0.031 0.031

242 0.009 0.008 0.008_——.. -. —-—

‘Date of Isotopic Analysis was 12/21/72.

of a practical capsule vent. Thus the releasecharacteristics of the fuel cylinder are important con-siderations for the safe containment of the fuelmaterial.

Investigation of helium release from actinide fuelsemployed in heat sources has produced a generaldescription of the sort of release behavior that mightbe expected from each fuel form. A study of releasefrom 2~Cm203 pellete10 showed that release wasquantitative as the helium was formed at atemperature of 1280”C. Release from 238PU02microsphere was found to be complete withinminutes at 1300°C, or above, at Oak Ridge NationalLaboratory (ONRL). 11 This was supported bybakeout studies on SNAP-27 fuel at MoundLaboratory12 and by microsphere studies atLASL.13114 The last work also examined the behaviorof several other *U02 fuel forms including cold-pressed and sintered cylinders (6.4 by 6.4 mm), solidsolution of plutonia with thoria and airconia, solidsolution cermets (SSC), plutonia-molybdenumcermets (PMC), and 51 -mm-diam by 6.4-mm-thickPMC disks. The cermets acted much like the

TABLE XXIX

SPECIFICATIONS AND PROPERTIES OF 3 3-W ‘%u0216 CYLINDERS

eeifbtions

Wt of fuel (g) —

Wt of -u (g) —

Power (W) 33.1 *0.2

Dimensions

Length (mm) 35.81 msxDiam (mm) 17.02 max

Cavity for PRDDiam (mm) 4.83 minDepth (mm) 6.60 min

Neutron emission rate

(n/s-kg fuel) x 1(T3 .-

(n/s-kg 23%%)x 103 .-

Qlinder SerialNumber35 36 37

73.545 73.506 74.07158.32’ 58.03= 58.39=33.042b 32.881b 33.085b

34.71 34.25 34.5916.71max 16.66max 16.64max16.61min 16.54min 16.56min

5.11 >4.83 >4.836.96 >6.60 >6.60

2695 2795 27153400 3540 3445

-——-.——---——‘Value based on calorimetry.

bWattage measured by Mound Laboratory (MRC).

21

TABLE XXX

CHEMICAL PURITY OF 29-W CYLINDERSSerial Numbers 38 through 41

releasing helium from the fuel under accident con-ditions.

B. Experimental

Element

LiBeBNaMgAlSiKCaTivCrMnFecoNiCuZn

w!??!

TABLE XXXI

PROPERTIES OF CONTROL PELLETS FOR 29-W CYLINDERS

SerialNumbers 38 through 41

8ampleNumber

487

Wt of pellet (g) 2.0045‘ Wt of -u (g) 1.376

DimensionsDiarn (mm) 0.808Height (mm) 0.394

Density (Mg fuel/m3 )a 9.94Percent of theoretical densityb 86.7

Neutron emission rate(n/s-kg fuel) x 1U3 2405(n/s-kg -u x 103 3500

—-—-..—.—..-————aCalculated from measured geometry and weight.

bBased on a theoretical density of 11.46 Mg/m3.

TABLE XXXII

OXYGEN-TO-METAL RATIO AND Hz OCONCENTRATION OF 29-W PELLETS

Oxygen/Metal 1.998 (*0.004)H20 Concentration (ppm)

g H20/1000-kg fuel)

TABLE XXXIII

PROPERTIES OF 29-W 235PU0216 HEAT SOURCESa

29 LPX3038 29LPX3039 29 LPX3040

Wt of heat source (g)Wt of fuel (g)

Wt of ‘% (g)b

Dimensions of heat source

o.d. (mm)

max

min

Length (mm)

max

min

Power (W)c

Specific power (kW/kg fuel)

Neutron emission rate(n/s-kg‘Pu) x 103

Dose rates in air at 10 cmGamma, R/h-kg 2%udNeutron, rem/h-kg%%

.------------- —____

40074.1850.98

32.5132.38

50.9350.7228.91

0.3897

3540

0.230.35

39974.0950.84

32.5132.21

50.83

50.7728.830.3891

3470

0.23

0.35

39974.1550.91

32.5132.23

50.7750.6528.87

0.3893

3570

0.220.36

aAllpropertiesweremeasuredon the fueledcapsuleas shownon IASL drawing 26Y-74997.bmsh ~Iues me based on calorimetry and Pu isotopic ~alYsis.

cPower measured calorimetrically.

‘Measured in air through 0.050*m Ta, 0.1 52-cm Ta-10W, and O.OSO-cmPt-20Rh.

direct current amplifier within the detection unit andadmission of the gas sample into the leak detectordirectly above the diffusion pump. The sensitivitynow extends over a range of about 2 x 10-8 to 2 x 10-2std cm3 He/s-g 238Pu02 16.

Still, the leak detector can be saturated withhelium. In that event the release is estimated bymeasuring the pressure within a known volume of thevacuum system with a Wallace Tiernan differentialpressure gauge, Model FA 141. This pressure is cor-rected for heating effects within the sample con-tainer. The Wallace Tieman gauge is calibratedagainst a mercury manometer.

2. Fuel Cylinder. The characteristics of the fuelcylinder used to date are shown in Table XXXV.Cylinder 5, shown in Fig. 18, was representative ofthe 30- W artificial heart fuel forml at the time thehelium release studies were initiated. Themicrostructure of this fuel form is given in Fig. 19.The fuel is of open porosity, which should permitrelatively free release of helium. Cylinder 5 was in

29 LLX3041

39974.1450.91

32.5132.23

50.7750.6528.87

0.3893

3510

0.250.35

29LPX3 047

39873.8350.45

32.5132.21

50.8550.7028.61

0.3875

3660

0.250.37

shelf storage for 857 days before initiating the steady-etate helium release experiment.

3. Procedure

a. Steady-Slate Testing. The fuel container wasconnected to the vacuum system, with both con-tainers having a helium atmosphere. The furnacewas open with water flowing through the coolingcoils and with a blower cooling the surface of the out-er container. Under these conditions the cylindertemperature was -350°C. The experiment was in-itiated by pulling a vacuum on the inner containerand allowing the fuel to self-heat to 552°C. Theblower was then stopped, the furnace was closed,and the cylinder was gradually heated to the 870”Coperating temperature. This temperature was main-tained throughout the experiment except whenequipment failure or maintenance work necessitateda shutdown. During the initial period of rapidlychanging release rate, the flow was monitored close-ly, but as the rate varied less and less, it wasmeasured twice a week.

.

.

.

24

TABLE XXXIV 10.0~ 1 I I I I I Ill 1 I I 1 I I I u

ISOTOPIC ANALYSIS OF 29-W CYLINDERSSerial Numbers 38, 39,40,41, and 47

Weight PercentPu Isotope (g Pu isotope/O.100-kgPU)

236 0.40 x 10+238 79.63239 16.39240 2.81241 0.83242 0.34

Fig. 8.Gamma spectrum of PNL 29-4 (energy in keV).

b. Temperature-Excursion Testing. ‘I’hetemperature excursion apparatus has just beenassembled and actual experimentation has not yetbegun. The planned procedure involves bringing thepellet used in steady-state testing to 870”C andholding it there until the helium release is stable:The temperature excursion will be to 1370”C over a45-rein interval.

a

C. Results*

The steady-state helium release of the ‘%%0 216cylinder 5 is shown in Fig. 20. The duration of the ex-periment was 5325 h, but downtime reduced the timeat temperature to 5041 h. Each interruption ofheating was characterized by a very low release rate.

1.0 —

(?

s

J!

2

0.10—

1

‘“m’loI [ 1 1 1 1 I I I 1 I I I # 1 1 J

100 1000

—

—

Olslanca (mm) From Csnter of Source

Fig. 9.TLD measurements on 29LLX3041.

la -fdt

Fig. 10.Sealed SO- G heat source.

- To shim

-~’oPuo:

-lo liner

_Ta-10W strengthmember

_Pt-20 Rh clad

2s

13.2

13.2

mm 1=-25.6 mm-i

.

..%7

klt

20”

1.52-MM well thickness

Fig. 11.Ta-10 W weld development capsule.

~32.51 .. —-

uelded In

~To-10W spring

— To liner

— To shim

_To-10W stm@hmombzr

—Pt-20 RtI clod

Fig. 12.Sealed 29- W heat source.

1mAfw320”Y-c’Weld #l‘“’ f”’”otitofl.o.

‘b

))7

Vents:

b 0.13.51

b=o.1’

Weld dlractlons are shown looking down cm capsule

CopmIla rotational speed = 127 mm/mln

Fig. 13.Ta-10 W 29- W welding diagram.

X 0.71 mm

X 0.71 mm

wide

wide

Y

\- Imw container238Pu O: cyllnder

Fig. 14.Vacuum system and equipment for He and Rnrelease measurements.

Resumption of heating produced somewhat morehelium after which the release rate returned to nearthe previously prevailing level within a few hours.

As can be seen, the fuel was brought totemperature over three working days. The large in-itial helium inventory recorded in Table XXXVproduced a large initial release rate that would haveoverwhelmed the leak detector if the heatup hadbeen any faster. After the fuel cylinder reached870”C, the release rate first decreased rapidly forabout 15 days, and then decreased more slowly. Therelease rate approached, but remained greater than,the helium generation rate for the fuel materialthroughout the observation period. It was still lso~.of the generation rate after -5000 h. This means that”the helium stored within the oxide cylinder con-tinually decreased during the experiment from an in-itial quantity of

.—

Fig. 16.Furnace and vacuum manifold assembly (insideglove box) for gas release experiments.

an important factor in determining the releasebehavior of this fuel form. The comparison withmicrosphere is uncertain because most data for thatfuel form were derived from short experiments(usually not exceeding 48 h). A comparison requiresextrapolation of microsphere release to longerperiods.

IX. RADON RELEASE FROM 238Pu0 216 FUELCYLINDERS

I 2 3456INI.(I. ,il:umt,. ~cltnli[ic l.:tlnarJlt,r>

cm of 1.( !JMIVt~$llV01 CALIFORNIA! 2, 3 4 5, ,6 7 8 9, 10, ,11 ,lJ 13.)4 )$,

..-— -—-—- 4

A. Introduction

Fig. 15.Welded nickel container for gas release ex-

periments.I

and the accumulation of additional helium with timein a plutonia cylinder and SSC and PMC fuel disksfollowing an initial “burst” of gas. 17,16 Cylinder 5apparently exhibited an intermediate type of releasedespite being a bulk fuel form like the cylinder anddisks mentioned above. This could be the result of theopen porosity of cylinder 5 (see Fig. 19). The poresmay have given rise to intermediate length diffusionpaths by which helium escaped from the fuel. Thelarge helium inventory of cylinder 5 may have been

Nominal 23*Pu embraces a number of plutoniumisotopes as shown in Table I. The decay chains ofseveral give rise to isotopes of radon. Thus ‘6Pu and240Pu eventually decay to zz~n, 23*PU and 24%u to

222Rn, and ‘@u to 21gRn. However, only the decayof 236Pu and 238Pu contribute significant radon overthe useful life span (~ 10 yr) of an artificial heartdevice. 19

At first consideration the zzzRn activity appears tobe insignificant when compared to that of 220Rn. Thelatter derived from 236Pu grows in rather rapidly. Tenyears after purification, a fuel which originally con-tained 1 ppm 236Pu; will have a 220Rn activity of

27

Fig. 17.Helium leak detection equipment.

5.76 x 105 dis/s/g 238Pu. In contrast, 222Rngenerated by the decay chain from 238Pu grows invery slowly due to the 248000 half-life of 234U. Thesame 10-yr-old fuel will possess a 222Rn activity of 1dis/s/g 238Pu. In a 30-W heat source this isequivalent to 60 dis/s.

But the amounts of the radon isotopes actuallyreleased by the fuel may be influenced by their half-lives. 220Rn has a short half-life of 51.5 s. Much of itmay decay before diffusing out of the fuel. Thelonger 3.8 day half-life for 222Rn presents thepossibility for a much higher escape probability.Thus the 222Rn activity become< important in anaged plutonium fuel and both radon isotopes must beconsidered in the design of a vented heat source cap-sule.

There is no existing experimental data pertinent tothe release of radon from actinide heat source fuels.However, some attention has been focused on theproblem. Parsont 20 has made calculations for the

SNAP- 19 heat source relating to the internalpressures attributable to radon and the biologicalrisks due to radon released from 238Pu0 216. He

28

TABLE XXXV

CHARACTERISTICS OF 238Pu0216 CYLINDERSerial Number 5

J&$?!Y

Date of Chemical Purification 9/28/69(date of electrorefining)

Date of Cylinder Sintering 11/25/69

Plutonium Isotopic Composition

@!!Y?fz Wt% Date of Analysis

236 0.70 x Id 12/8169

238 79.82 12/29/69

239 16.21 12129169

240 2.92 12f29169

241 0.80 12129/69

242 0.25 12129169

Physical Properties (1 1/25/69)

.Diam (mm)Height (mm)Weight (g)Weight of Pu (g)Weight of ‘% (g)Density (Mg fuel/m3)YO of theoretical densityPower (W)Power density MW/m3

Helium Content

857 days after sintering, cm3

21.5420.4073.6664.9251.8210.1188.229.8

4.03

89.94

concluded that the pressures would be negligibleand that the potential biological effect would besmall. However, the latter conclusion was based onseveral assumptions that are not appropriate for anartificial heart heat source. The objective of the pre-sent work, therefore, is to determine experimentallythe portions of the radon isotopes that are releasedfrom a238Pu0216 cylinder.

.

.

.

●

. . I

Fig. 18.238PU0216 cylinder, serial number 5 (scale in

inches).

B. Experimental

1. Apparatus. The vacuum system and equip-ment used in sampling radon release from theceramic fuel body at the operating temperature ispresented in Fig. 14 and has been discussed in Sec.VIII.B. 1. (a). The liquid nitrogen-cooled radon trapsshown in Fig. 21 were located outside the glove box,although the vacuum lines returned into the box.Again fuel integrity, rather than a filter in thevacuum line, was relied upon to prevent severe 23*Pucontamination of the cold traps. A filter could havedelayed the migration of the radon gas to the traplong enough to allow complete decay before enter-ing the trap.

* The cold traps were placed in parallel but atdifferent distances from the oxide pellet. Collectingsamples at two. locations would make it possible to.calculate the amount of decay occurring betweenthe pellet and the traps. With a correction for decay itwould then be possible to determine the amount ofradon that escaped through the surface of an oxidecylinder.

. . .

Fig. 19.Typical microstructure of 238Pu0216 cylinder (SQOX magnification).

The identification and determination of radonreleased from the fuel is achieved by alpha countingand gamma spectroscopy.

2. F’uel Cylinder. The cylinder employed in theradon release experiment was described undermaterial for helium release. This fuel had the advan-tages of being well characterized (see Table XXXV)and being a representative material with a knownthermal history. It also had attained an age (morethan 900 days since chemical purification) commen-surate with an abundance of 22~n.21 It was notsufficiently aged, however, to contain a detectablequantity of 222Rn.lg

3. Procedure. Sampling occurred, after the fuelhad reached operating temperature, by evacuatingthe inner capsule through one of the cold trapsshown in Fig. 22. An examination of the 236Pu decay

29

11 I‘6 717°C:O1o

;g.’w 870”C

o

j{ ,.-7.

$ ~?

[1

1

-.--—.

E.

.= — ——-8

—:;10 Helium generation rote

z( 1.63 x10-8std cm3/g238Pu0’~ s

IO”9J 1, I40 80 120 160 200 240 280

Time (days)

Fig. 20.Helium release rate from 238Pu0 z‘6 cylinder, S.

series shown in Eq. (1) indicated that the short half-lives of ‘ORn and 21@o would cause them to decaybefore the sample could be counted, but that 21@bshould be ideal for detection purposes, Equation (2)reflected the relationships prevailing in the systemprovided that the actual -n release rate wasconstant with time or at least showed no trend duringsample collection. Since R ~a20R”wasconstant over thecollection period, the formation of 21@b representedan instance of approach to secular equilibrium. As acompromise between largest possible sample size

and greatest possible accuracy in Rpalzpb the samplewas collected over approximately two half-lives of212Pb.

Fig. 21.Radon traps and associated vacuutn system exter-nal to glove box enclosure.

(1)

236 Pu 2.85 y 3.64d232U 72y , zz6Th 1.91y , 224Ra—>●

a (Y a a’

51.5 s220R,, _, 0.16s216P0 ● 10.64 h212pb ● 60.6 m212Bi ● ‘T1a a B a (34%)

P(66%)

II

60.6 m 8 3m

2,2P0 0.3p s , 208~ba (stable)

The sample was then isolated from the vacuum The measured 212Pb activity was corrected for the

system while noting the duration of the sample decay that occurred during the interval between

collection period. After the ‘ORn had completely trapping termination and activity determination. The

decayed, the mqority of the contents was diesolved activity was converted to 212Pb abundance in the trap

out of the trap with 5 ml of concentrated HNOs at termination. The RpZi2ebin Eq. (2) was calculatedfollowed by 5 ml of agua regia. The solutions were from Eq. (3), which assumes that the abundance ofcombined and were subjected to radiochemical 212Pb in the trap at t = O was zero.analysis.

30

.

.

.

●

I

.

.

44’zlzBi, and Z08TI(au 220Rn daughters) as well as ‘%,

The present procedure did not permit detection of222Rn. The 212Pb activity corresponded to R #Rn of-86 atoms/s, which is about 1.5x 10-3% of the 220Rngenerated per second. Rr~20Rn is proportional toRtzz.m ,additional experiments are required to deter-

+ “-”--~~ ~~e:n~~~easured. Thus the

The 2*u activity in the sample was 44 dis/s. Theassociated 236Pu would weld a 212Pb activity of only

.

.

Fig. 22.Glass reservoir for trapping Rn.

R212Pb = RP216P0 = Rt220Rn @Rr220Rn fl Rg220Rn , (2)P

where RP = rate of production within the trap, R t =rate of trapping, R, = rate of release from the pellet,and ~ = rate of generation within the pellet.

Rp212pb = AzN*/(1 –e-izt) , (3)

where )i2 = decay constant for 21@b, in S-l, N2 =number of atoms of 212Pb at termination, and t =duration of production or trapping in seconds.

C. Results

The sample was collected over 20.5 h. Thea and -yspectrums revealed the presence of 21@b, 21*0,

measured 212Pb activity ‘was the result of ‘220Rnrelease and not the result of contamination of thesample by plutonia.

The radon release experiments are a continuingproject. Future experiments are designed to deter-mine release rates from the fuel cylinder.

X. FUEL CHARACTERIZATION

A. Nondestructive Analysis

Emphasis has been placed on developing non-destructive methods for characterizing medical-grade 238Pu fuels. At present, the most valuablesingle measurement to ascertain fuel radiation quali-ty is the neutron count. This measurement issupplemented by detailed gamma-ray spectroscopy,which is capable of detecting small amounts of lightelement impurities, such as fluorine, sodium,chlorine, nitrogen, oxygen, phosphorus, andaluminum. Gamma-ray spectroscopy is also used forroutine measurements of 170 and 180 in 2WPU0 216.

1. Neutron Counting. Neutron counting ofsmall samples is done routinely inside a well sur-rounded by four detectors, but large samples must becounted on top of the well about 42 cm above the in-well position and 20 cm above the top of themoderator. Counting inside the well is much less sen-sitive to sample positioning, however, than countingoutside the well. Quantitative measurements showedO.1% variation of count rate for each millimeterdifference in height inside the well as compared witha 2% variation above the well. These results point tothe importance for accurate positioning of the sampleand standard for above-well counting, and for thedesirability of constructing a new well counter tohandle large samples. Development work is also be-ing continued on the “long counter” detectionsystem. It consists of five matched 3He detectors in a

I

polyethylene , moderator arranged to achieve asimilar response to a wide variety of initial neutronenergies.

2. Gamma Spectrometrya. Measurement of’70 and ’80 in ‘38Pu0 z ‘6. ‘I’he

reliability of gamma spectrometry for measuring ’70and 180 in ‘*u0216 was estimated from computer-compiled spectra to be + 10°\ofor 180 contents downto 10 pg/g total oxygen and +25Y. for 170 contentsdown to 50 ~g/g of total oxygen. This estimate wasbased on measurements made by compiling a seriesof gamma spectra with a PDP- 11 computer fromdefinite ratios of a 23%k02nd and 238Pu metalspectrum. This was done by accumulating gammaspectra of the 23%%02nat (2040 ppm 180 and 370 ppm170) and for 238Pu metal. Successive additions ofdata from the latter spectrum to that of the formerproduced a series of new spectra with 180 and 170abundances effectively reduced by half each time.Channel-by-channel addition of the spectral datawas done on a PDP- 11 minicomputer. From thisseries of spectra we concluded that gamma-rayanalysis is dependable for determining lUOabundances down to less than 10 ppm. The estimatedaccuracy in the 10- to 2040-ppm range is about 10%.Calculations of 170 abundances are unreliablebelow the 50-ppm level. Poor accuracy is due mainlyto low statistics in counting the peaks of interest. Theaccuracy is estimated to be 25’%.

b. Determina~ion of 228Th. The measurement of228Th in ‘6Pu sources by gamma spectrometry wasinvestigated by comparing the intensity of the 861-keV peak, from the 208Tl daughter of 228Th, to that ofthe 851-keV peak of 238Pu in aged ( >28 days)sources. Results for five samples showed a highpositive bias relative to those obtained with achemical method for determining 228Th content. Apossible reason for the high values obtained by gam-ma spectrometry may be that 220Rn (54.5-s half-life)

‘ diffuses toward the surface of the sample. The 20~1activity resulting from the diffused ‘ORn would besubject to less attenuation by the sample than the238Pu activity to which it is compared. If thismechanism proves valid, then gamma dose rates, animportant specification for medical-grade 23*u,could be greater than calculated from original236Pu content. It seems likely that gammaspectrometry is not a reliable method for the non-destructive determination of 23~h content.

c. Minicomputer Data Reduction of Gamma-RaySpec[ra. A DPD-11 computer with 8 K of corememory has been interfaced with the 4096-channel

analyzer for measuring gamma-ray spectra.Associated software provides a means of data reduc-tion, including the calculation of gamma-rayenergies and peak area. A first application may in-volve the determination of 345- and 351 -keV peakareas for use in 180 analysis in 238Pu. Preliminaryresults indicate a linear relationship of calculated351-/345-keV peak ratios with 180 content, but thelinearity does not hold true when the spectrum shiftsslightly. We observed this effect in spectra preparedby combining a spectrum of Pu02nat (2040 ppm 180)with that of plutonium metal (O ppm 180) in suitableproportions by using the computer. Thus, equalproportions of each individual spectra produce aneffective standard containing 1020 ppm O. Two

“distinct linear segments, separated by a break, wereobtained for the relationship of 351 -/345-keV peakarea ratios with 180 content, as calculated by thecomputer. Examination of the computer program dis-closed the fact that shifting the peak channel, upon

“which the peak area is based, caused the break in thecurve. Efforts are being made to modify the programso that there will be less dependence of peak area topeak channel.

3. Calorimetry. An evaluation of LASL’S waterbath calorimeter was made using a 30-W 23@u0 216heat source, 30ULXOO07, which had been standar-dized by Mound Laboratory (MRC). The standar-dization was done in a high-precision isothermaltwin-bath calorimeter, calorimeter No. 87. MRCprovided wattage tables (i.e., wattage as a function oftime), which are accurate to a least +0.04~0 at the95% confidence level, for the standardized source.

Over a one month period, 15 determinations weremade of the wattage. These results, together with theMRC values, are given in Table XXXVI. Eight LASLresults were higher than the MRC values with anaverage difference of 0.024 W and seven were lowerwith an average difference of 0.022 W. The averagedifference between the two sets of values is +0.023 Wor +0.08Y0. Thus the performance of the LASL waterbath calorimeter is within its design criterion of+0.lYo at 30 W.

B. Chemical Analysis

1. Mass Spectrometric Analysis of 170 and l@in 238Pu02 16. Since a low neutron emission rate isessential for medical-grade fuels, measuring tracesof 180 and 170 in 23~u0 z16 is necessary becausethese impurities contribute significantly through

.

.

.

32

TABLE XXXVI sequent analyses. The furnace tube, which was

.

.

.

coated with carbon and sample on the ineide, wasCOMPARISON OF IASL AND MRC CALORIMETRIC

MEASUREMENTS ON %S0216 HEAT SOURCE 30ULXOO07replaced with a new fused-silica tube. Further deter-minations were planned to estimate the nrecision ofthe method. -

.

Det

No.—

123456789101112131415

Date

s- 5-71

s- 671

8-10-71

8-11-71

8-16-71

8-1 S-71

8-19-71

8-23-71

8-24-71

8-25-71

8-26-71

8-27-71

8-30-71

8-31-71

9- 1-71

LASL Value

(w)

28.917

28.935

28.972

28.901

29.012

28.943

28.930

28.926

28.893

28.910

28.910

28.931

28.S67

28.904

28.937

MRC Vdu+

(w)

28.93228.93228.92928.92928.92628.92528.92428.92228.92128.92128.92028.92028.91828.91728.916

Difference

(LASI/MRC)

(w

-0.015+0.003+0.043-0.028+0.086+0.01s+0.006+0.004-0.028-0.011-0.010+0.011-0.051-0.013+’0.021

aValues from MRC’S Calorimetric Standardization Calendar for30ULXOO07.

(a,n) reactions to the neukon count. Absolutemeasurement of these isotopes by coupling inert gasfusion and mass spectrometry is being studied. Theoxide sample is heated to 2000°C in a helium at-mosphere with elemental 1~ to produce CO which

is collected in a silica gel trap and then anal zed ue-3ing a mass spectrometer.The ratio of the l% 70, 30,

peak to the 13C160, 29, peak is used to determine the1’70/160 atom ratio and the ratio of l~leO, 31, peakto the 13C160, 29, peak is used to determine the180/160 atom ratio.

In initial tests, a tungsten crucible was used to con-tain the oxide sample while it was heated, but thiswas replaced by a tantalum crucible to avoid thevolatilization of tungsten oxides. Attempts to pelletize

23*u02 mixed witha sample of 13C before it washeated in the analysie apparatus to 2000°C was notpractical because of the poor pelletizingcharacteristics of the 13C.

Analyses of powdered 238pu02.13C mixtures

ghowed the precision of the method to be unsatisfac-tory. Disassembly and inspection of the analysis ap-paratus indicated that significant amounts ofpowdered samples were blown out of the crucibleand adhered to the wallE of the fused silica furnace

2. Measurement of O/M Ratios in238PU02. The reliability of the thermogravimetricmeasurement of oxygen-to-metal atom ratios (0/M)in 23@u02 fuels was investigated using oxides O:known composition prepared from 23*u metal (80%enriched in 23*Pu isotope). In this method, originallydeveloped for analyzing 23~u02 materials, theoxide is heated to 1000° C in air to form ahyperstoichiometric oxide which is then reduced inAr-6Y. H2 at 1000”C to form the stoichiometricdioxide.22 The change in weight is used to calculatethe original O/M. In these tests the ‘*u metal wasfirst carefully weighed in an inert atmosphere andthen slowly converted to a hyperstoichiometric oxideby heating in air. Weighed portions of this oxide con-taining known amounts of 2WPU were oxidized andthen reduced as described above. Twelve deter-minations showed that the standard deviation inmeasuring O/M in ‘*u02 was +0.004. Studies ofthe effect of uranium impurity on the measurement o{O/M are planned.

REFERENCES

1. “Preparation and Evaluation of Medical-GradePlutonium-238 Fuels, July 1, 1967- June 30, 1971,”Los Alamos Scientific Laboratory report LA-4940(1972).

2. D. A. Hicks, J. Ise, Jr., and R. V. Pyle,“Probabilities of Prompt-Neutron Emission fromSpontaneous Fiseion,” Phys. Rev. 101, 1016 (1956).

3. V. A. Druin, V. P. Perelygin, and G. I. Khleb-nikou, “Spontaneous Fission Periods of 237Np, 23~U,and 242Pu,” Soviet Phys. JETP 13,5,913 (1861).

4. L. J. Mulline, G. M. Matlack, J, Bubemak, and J.A. Leary, “Characterization and Properties ofMedical-Grade 2%Pu Fuels,” Second InternationalSymposium on Power from Radioisotopes, Madrid,May 29- June 1, 1972, OCDE, AEN, 49-67.

tube during analysis. To avoid sample loss, contain- 5. J. A. Porter and M. C. Thompson, “Preparationment of powdered ‘Pu02 - 13C mixtures in smalltantalum envelopes was tried successfully. However,

of 238Pu02 16,” Savannah River Laboratory reportDP-1153 (1968),

a high apparatus background developed upon sub-

33

6. “PreP&ation and Evaluation of Plutonium-238Radioisotopic Heat Sources, Annual Report for thePeriod July 1, 1868- June 30, 1969,” LASL internaldocument (1970).

7. M. Taherzadeh, “Neutron Yield from the (a,n)Reaction in the Isotope Oxygen-18,” Nucl. Sci. Eng.44, 190-193 (1971).

8. I. A. Serdivkova, A. G. Khabakhp